This invention generally relates to leak testing and more particularly relates to a system and method for in situ testing of the leak-tightness and leak rate of a tubular member, which tubular member may be a heat transfer tube of the kind found in typical nuclear steam generators.
Although leak testing devices and methods are known, it has been observed that these devices and methods have a number of operational problems associated with them which make such devices and methods less than completely satisfactory for leak testing heat transfer tubes of the kind found in typical nuclear steam generators. However, before these problems can be appreciated, some background is desirable as to the structure and operation of a typical nuclear steam generator.
In this regard, a typical nuclear steam generator or heat exchanger generates steam when heat is transferred from a heated and radioactive primary fluid to a non-radioactive secondary fluid of lower temperature. The primary fluid flows through a plurality of U-shaped tubes disposed in the steam generator. The secondary fluid flows across the exterior surfaces of the tubes as the primary fluid flows through the tubes. The walls of the tubes function as heat conductors for transferring the heat from the heated primary fluid flowing through the tubes to the secondary fluid of lower temperature flowing across the exterior surfaces of the tubes. As the heat is transferred from the primary fluid to the secondary fluid, a portion of the secondary fluid vaporizes to steam for generating electricity in a manner well understood in the art.
Occasionally, due to tube wall cracking caused by stress and corrosion during operation (i.e., known in the art as primary water stress corrosion cracking), the steam generator tubes may degrade (i.e., experience tube wall thinning) and thus may not remain leak-tight. If through-wall cracking occurs due to the degradation, the radioactive primary fluid may undesirably leak through the crack and commingle with the nonradioactive secondary fluid.
Therefore, such leaking tubes may be plugged to prevent commingling the radioactive primary fluid with the nonradioactive secondary fluid. However, for safety reasons, technical specifications imposed by the United States Nuclear Regulatory Commission on each nuclear reactor power plant holding an operating license set a limit on the percentage of heat transfer tubes that may be plugged. Exceeding this technical specification limit jeopardizes the operating license of the power plant. It is therefore prudent to plug only those heat transfer tubes that require plugging.
However, it has been observed that not all leaking tubes require plugging because a small amount of tube leakage is tolerable. This is so because small amounts of tube leakage do not pose safety concerns. In view of this, the United States Nuclear Regulatory Commission has issued Draft Regulatory Guide 1.121 (NUREG-1477) which provides a basis for determining acceptable tube leakage in order to identify which degraded or leaking steam generator tubes must be plugged in order to satisfy safety requirements. In general, Draft Regulatory Guide 1.121 sets a limit for an acceptable leak rate at a specified pressure in the tube, the pressure being that which would be experienced during a postulated accident (e.g., main steam line break). Therefore, under the requirements of Draft Regulatory Guide 1.121, a degraded or leaking tube maybe "leak-rate tested" to determine whether or not the tube will have an acceptable leak rate during such a postulated accident. If the leak-rate is acceptable, then the leaking tube need not be plugged. Of course, it is desirable to perform such a leak-rate test in a precise and cost efficient manner.
Techniques for determining the leak rate from degraded nuclear steam generator tubes are known. In this regard, it is known that as an eddy current inspection probe is translated along the inner diameter of a tube, its voltage amplitude will increase when it passes a degraded portion of the tube. The degraded portion of the tube may have a through-wall crack which will allow fluid to leak therethrough at a flow rate related to the size of the crack and the fluid pressure within the tube. That is, the voltage amplitude will have a statistical correspondence with the flow rate through the crack at a given pressure. Thus, by knowing the voltage amplitude and internal tube pressure, one can determine the flow rate through the crack. In other words, the voltage amplitude of the eddy current probe can be correlated to a given flow rate and pressure. It is therefore possible to make a plurality of empirical or experimental measurements of voltage, leak-rate and pressure to establish a data base of flow rate as a function of eddy current voltage and anticipated pressures. Such a database can then be used in the field to conveniently determine the anticipated leak rate at the pressure of interest merely by measuring the voltage amplitude, as explained more fully hereinbelow. Typical prior art methods of gathering the empirical data to establish the data base are discussed hereinbelow.
It is important to precisely and cost-effectively determine the empirical data for the data base. However, two prior art methods typically used to obtain the empirical data for the previously mentioned data base are costly and may be imprecise. One of these prior art methods entails constructing a bench-scale model steam generator and then inducing through-wall degradation in the tubes that are disposed in the model steam generator. Eddy current voltage amplitude measurements of the degradation are then made by passing an eddy current probe through the tube. Next, the degraded tube is pressurized at a plurality of pressure values and the pressures and leak rates measured. The eddy current voltage amplitude, pressures and leak rates are recorded in a data base. However, a problem with this method is that the bench-scale model steam generator is constructed at a not insignificant expense and the empirical data obtained may or may not be truly representative of the real conditions in a full-sized steam generator belonging to an actual operating nuclear power plant.
Another prior art method of establishing the data base entails inserting an eddy current probe into a degraded tube of a full-sized steam generator belonging to an actual power plant to obtain the required voltage reading and then removing the tube from the steam generator. Once the tube is removed, the tube is leak-rate tested in order to establish the previously mentioned leak rate and pressure values for the data base. However, applicants have observed that the process of removing the tube to be examined may result in a change in morphology of the degraded portion of the tube such that the morphology of the degraded portion of the tube after removal is not the same as before the tube is removed. This may occur, for example, when a network of cracks in the degraded portion of the tube are connected by ligaments which become torn during the removal process. Thus, the morphology of the tube after removal may be different than before removal. Applicants have confirmed this phenomenon by observing that eddy current voltage amplitudes are sometimes larger after removal than before removal. Therefore, this second prior art method of establishing the data base may cause the data base to be imprecise and is therefore not preferred. In addition, this second prior art method of establishing the data base cannot be conveniently performed in situ; that is, the tube must be removed. Moreover, the process of removing the tube is time consuming and necessitates that the steam generator betaken off-line for an extended period of time. Each day that the steam generator is off-line requires the reactor owner to incur approximately $1,000,000 in replacement power costs.
Therefore, a problem in the art is to provide a technique for steam generator tube in situ leak-rate testing that is precise and cost effective in order to establish the data base while overcoming the disadvantages of the prior art.
After the data base is established, an eddy current measurement is performed in the field on a preselected steam generator tube to obtain the voltage amplitude measurement. Next, the voltage amplitude is looked-up or found in the data base to obtain an anticipated leak flow rate at an anticipated pressure (e.g., pressure during a postulated accident). In this manner, the flow rate through a leaking steam generator tube of an operating power plant during a postulated accident may be determined simply by performing an eddy current inspection of the tube and then correlating that eddy current voltage amplitude to flow rate and pressure data existing in the data base. However, in order to establish the values for the data base, a plurality of steam generator tubes first must be leak-rate tested.
Yet another apparatus and method for testing the tightness of closed-end tubes in heat exchangers of nuclear reactors is disclosed in U.S. Pat. No. 3,919,880 titled "Method And Apparatus For Testing Closed-End Tubes In Heat Exchangers of Nuclear Reactors And The Like" issued Nov. 18, 1975 in the name of Gunter Seyd, et al. This patent discloses a method and apparatus for detecting a leak in a heat exchanger tube of a nuclear reactor by inserting an expandable plug which seals off a closed region in the heat exchanger tube. A pressurized fluid is then introduced into the closed region, and the pressure of the fluid is monitored for detecting a leak through the closed end of the tube. However, this patent merely discloses testing for the leak-tightness of a closed-end tube with a fluid under pressure, and merely discloses sealing the tube at one point with an expandable plug and testing the area located between the expandable plug and the closed end of the tube to determine if the closed end of the tube permits leaking. This patent does not appear to disclose leak testing for fluid flow rate through a crack in the walls of an open-ended tube. Leak testing of an open-ended tube is required for leak testing steam generator heat transfer tubes in the manner satisfying Draft Regulatory Guide 1.121 (NUREG-1477).
Therefore, what is needed are a suitable system and method for in situ testing of the leak-tightness of a tubular member, which tubular member may be a heat transfer tube of the kind found in typical nuclear steam generators.